ML19326C004

Cycle 2 Reload Rept. Nonproprietary Version

ML19326C004
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/30/1976
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1433, BAW-1433-01, BAW-1433-1, NUDOCS 8004180696
Download: ML19326C004 (18)

Text

. . - _ _ _ _

BAW-1433 November 1976 ARKANSAS NUCLEAR ONE - UNIT 1. CYCLE 2 4

- Reload Report -

O I

)

O g004180 hh b d O .

Babcock &Wilcox

,ps, 1

i t

i ARKANSAS NUCLEAR ONE - UNIT 1, CYCLE 2

- Reload Report -

~

l I

i t

1

}

I i

l I,

L t

j l

l

'1 I

! BABCOCK & WILCOX

! Power Generation Group i Nuclear Power Generation Division 7 P. O. Box 1260 1 Lynchburg, Virginia 24505 l

1 i

Babcock & Wilcox '

o - = - - _ _ _ _ - _ . . . . . . . . , _ _ _ . _ _ _

h 4

( ,-

BAW-1433 November 1976 l

i f

ARKANSAS NUCLEAR ONE - UNIT 1, CYCLE 2 j

- Reload Report -

4 I

4 i

I l

f 1

I r

BABC0CK & WILCOX Power Generation Group Nucicar Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox

[

n U l. INTRODUCTION AND

SUMMARY

This report justifies the operation of the second cycle of Arkansas Nuclear One (ANO), Unit 1 at the rated core power of 2568 MWt. Included are the re-quired analyses outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," dated June 1975. Supporting Cycle 2 opera-tion of ANO Unit 1, this report employs analytical techniques and design bases established in reports that have been accepted by the USNRC (see references).

Cycle 1 and 2 reactor parameters related to power capability are summarized briefly in section 5. All accidents analyzed in the FSAR have been reviewed for Cycle 2 operation. In cases where Cycle 2 characteristics were conserva-tive as compared to those analyzed for Cycle 1 operation, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required f%

for Cycle 2 operation are justified in this report. Based on the analyses and considering the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, ANO Unit 1 can be operated safely during Cycle 2 at the rated core power level of 2568 MWt.

l l

4 1

. s_-

Q)'

1-1 Babcock & Wilcox i

I I

C 0 q

b 2. OPERATING HISTORY Unit 1 of the Arkansas Nucleas: One (ANO) station achieved initial criticality on August 6, 1974, and power escalation comme.ced on August 13, 1974. The 100% power level of 2568 MWe was reached on 7ecember 8, 1974 A control rod interchange was performed at 242 effective 'ull-power days (EFPD). The design fuel cycle is scheduled for completion in January 1977, af ter 490 t 10 EFPD.

The first cycle involved no operating anomalies that would adversely affect ,

fuel performance during the second cycle.

Operation of Cycle 2 is scheduled to begin in March 1977. The design cycle length is 272 1 10 EFPD; no control rod interchanges are planned.

O

, s 2-1 Babcock & Wilcox

(~ ,3

~

3. GENERAL DESCRIPTION The Arkansas Nuclear One (ANO) -- Unit 1 reactor core is described in detail in Chapter 3 of the FSAR.I The Cycle 2 core consists of 177 fuel assemblies (FA),

each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. The fuel rod cladding is cold-worked Zircaloy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch. The fuel consists of dished end, chamfered, cylindrical pellets of ura-nium dioxide which are 0.700 inch in length for batches 2 and 3, 0.600 inch '

long for batch 4 and 0.370 inch in diameter. (See Tables 4-1 and 4-2 for addi-tional data.) All FAs in Cycle 2 maintain a constant nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths and theoreti-cal densities vary slightly between batches. Specific values are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for ANO-1, Cycle 2. The initial enrich-(N ments of batches 2 and 3 were 2.72 and 3.05 wt % uranium-235, respectively.

Batch 4 is enriched to 2.64 wt % uranium-235. All the batch 1 assemblies will be discharged at the end of Cycle 1. Some of the batch 2 and 3 assemblies will be shuffled to new locations. The batch 4 assemblies will occupy primarily the periphery of the core and eight locations in its interior. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of Cycle 2.

Reactivity control is supplied by 61 full-length Ag-In-Cd control rods and soluble boron shim. Besides the full-length control rods, eight partial-length axial power shaping rods (APSRs) are provided for additional control of axial power distribution. The Cycle 2 locations of the 69 control rods and the group destunations are shown in Figure 3-3. The core locations of the total pattern (69 controls rods) for Cycle 2 are identical to those of the reference cycle indicated in Chapter 3 of the FSAR.I However, the group designations differ between Cycle 2 and the reference cycle to minimize power peaking. Neither l

! control. rod interchanges nor burnable poison rods are required for the de-s signed operation in Cycle 2.

s .

('") l l 3-1 Babcock & \Milcox i

l l

mm. m

r- m Tha nominsi cyc.Se precaura is 2200 psia, and the core average densified nomi-nal heat rate is 5.81 kW/ft at the rated core power of 2568 MWt. There were s no relevant changes in core design between the reference and reload cycles.

, ) The same calculational methods and design information were used to obtain the important nuclear design parameters. In addition, no significant operational procedure changes exist from the reference cycle with regard to axial or radial power shape control, xenon control, or tilt control. The operational limits ,

(Technical Specifications changes) for the reload cycle are shown in section 8.

A fuel melt limit of 19.4 kW/ft has been employed in calculating the reactor protection system (RPS) setpoints. Thirteen ANO-1 batch 3 FAs have centerline melt limits between 19.2 and 19.6 kW/ft (class 3 fuel). However, only two as-semblies, IC56 and 1C59, need selective loading based on 19.4 kW/ft. Thus,.a sufficient fuel melt margin will be maintained through cycle 2.

O 1

1 i

l

((() l l

3-2 Babcock & \Milcox l l

l

p ^ i Figu.ee 3-1. Core Loading Dia sam for ANO Unit 1, Cycle 2 l

Fuel Transfer Canal D 4 4 4 4 4 4 g 4 4 4 2 3 2 4 4 4 F-7 A-8 F-9 y 4 3 A-6 3 2 C-6 2

D-5 3

B-8 2

D-11 2

C-10 3

A-10 3

8-12 4

B-4 p 4 3 2 3 3 4 3 ,4 3 3 2 3 4 D-2 L-8 A-7 B-5 C-4 9-11 A-9 51 - 1 1 D-14 g 4 3 3 3 2 2 2 2- 2 3 3 3 4 F-1 U-1 C-3 E-6 B-7 C-4 C-9 E-10 C-13 C-15 F-15 F 4 4 2 3 '2 2 3 2 3 2 2 3 2 4 4 F-3 E-2 F-5 D-7 B-6 C-8 B-10 C-12 F-11 E-14 F-13 i i

6 4 2 2 4 2 3 3 2 3 3 2 4 2 2 4 C-6 E-4 G-2 F-2 D-3 C-8 C-12 F-14 C-14 E-12 C-10

=== _

E 4 3 3 3 2 2 2 2 2 2 2 3 3 3 -4 I 11 - 1 ti-2 N-? N-7 H-3 it-7 H-8 H-9 H-13 D-9 D-13 H-14 M-15

==- __

g 4 2 2 4 2 3 3 2 3 3 2 4 2 2 4

. K-6 M- K-2 L-2 tb4 K-8 N-13 L-14 K-14 l1-12 K-10 j-g 4 4 . 3 2 2 3 2 3 2 2 3 2 4 4 L-3 II-2 t-5 K-4 P-6 0-8 P-10 N-9 L-11 M -la L-13 N 4 3 3 3 j 2 2 2 2 2 3 3 3 4 L-1 K-1 0-3 M-6 P-7 K-12 P-9 M-10 0-13 K-13 L-13 4

y , 4 3 2 3 3 4 3 4 3 3 2 3 4 N-2 ti-5 R-7 ' P- 5 0-12 P-11 R-9  !!-8 N-14 0 . 4 3 3 2 2 3 2 2 3 3 4 P-4 R-6 0-6 N-5 P-8 N-11 0-10 R-10 P-12 P 4 4 4 2 3 2 4 4 4

• L-7 R-9 L-9 R 4 4 4 4 4 (

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Batch Previous Core Locatiort 3-3 Babcock s.Wilcox

Figure 3-2. ANO Unit 1 Enrichmsnt ana Burnup Distribution for Cycle 2 8 9 10 11 12 13 14 15 O

2.72 2.72 2.72 2.72 3.05 3.05 3.05 2.64 14.316 18,613 18,165 18,814 14,083 16,160 12,140 0 W

3.05 3.05 2.72 2.64 2.72 2.72 2.64 K

13,170 13,872 15,368 0 16,618 19,742 0 2.72 2.72 3.05 2.72 2.64 2.64 13,628 18,835 12,608 16,493 0 0 3.05 3.05 3.05 2.64 i 9,351 11,616 8,944 0 N

2.72 3.05 2.64 ,

19,794 3,743 0 2.64 0

P l

R 2.72 Initial enrichment 14,316 BOC burnup; Wd/mtU l

l lO I

3-4 Babcock & Wilcox l

l f

l

_ . . ~ _

Figura 3-3. Control Rod Locations for ANO Unit 1, Cycle 0 X

A 3 4 5 4 C 2 7 7 2 D 6 8 3 8 6 E 2 3 1 l' 3 2 F 4 8 5 6 5, 8 4 C ' 1 3 3 1 7 W- 5 3 6 4 6 3 5 -Y K 7 1 3 3 1 7 L 4 8 5 6 5 8 4 H 2 3 1 1 3 2 N 6 8 3 8 6 O 2 7 7 2 P 4 5 4 R

l.

z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Croup Number of Rods Funetten 1 8 Safety

'X Croup Number 2 8 Safety 3 12 Safety 4 9 Safety 5 8 Control 6 8 Control 7 8 Control 8 8 APSRs 1

TOTAL 69 V

3-3 Babcock & Wilcox

( m l :/ 3 U 4.

FUEL SYSTEM DESIGN 4.1.

Fuel Assembly Mechanical Design Pertinent fuel design paran.eters are listed in Table 4-2. All fuel assemblies are identical in concept and are mechanically interchangeable. The new FAs have modified end fittings, primarily to reduce FA pressure drop and to in-crease holddown margin. All other results presented in the FSAR fuel assembly mechanical discussion are applicable to the reload fuel assemblies.

4.2. Fuel Rod Design Pertinent fuel rod dimensions for residual and new fuel are listed in Table 4-3. The mechanical evaluation of the fuel rod is discussed below.

4.2.1. Cladding Collapse Creep collapse analyses were perforned for three cycle assembly power histories for ANO-1 and yielded acceptable results. A three-cycle design power history for batch 4 fuel has not been finalized; thus, no collapse analysis was per-i formed for this batch. However, batch 2 and 3 fuel is the worst case for Cycle 2 operation due to its prevfous incore exposure time. Since the analytical parameters of the fresh batch 4 fuel are similar to those of the limiting batch 3 fuel, and batch 4 will have a shorter three-cycle exposure, there is no col-lapse concern wf:S batch 4 fuel. Batch 2 and 3 fuel pellets were resintered, and the lowest (woz st-case) density was determined from as-fabricated data.

This was coupled witn the projected worst-case assembly power history to deter-mine the most limiting collapse time as described in BAW-10084P-A.2 Measured power distribution data obtained during Cycle 1 oaeration confirmed the accuracy of the Cycle 1 design calculations used for the collapse analysis. The conserva-

tisms of the analytical procedure are summarized below.

-1. The CROV computer code was used to predict the time to collapse. CROV

' conservatively predicts collapse times, as demonstrated in reference 2.

2., Ne credit is taken for fission gas release. Therefore, the net differen-t tial pressures used in the analysis are conservatively high.

4-1 Babcock & Wilcox

e

3. Tha cledding thickn:co uccd w s tha icwxr toleranca linit (LTL) of tha es-built measurements. The initial ovality of the cladding was the upper

, ,x tolerance limit (UTL) of the as-built measurements. These values were

() taken from a statistical sampling of the cladding.

4. Batch 2 and 3 fuel pellets underwent additional sintering, producing a range of densities. For this collapse analysis, the rods that had the lowest as-fabricated pellet densities were conservai.1vely assumed to be located in the worst-case power region of the core.

The most limiting assembly was found to have a collapse time greater than the maximum orojected three-cycle exposure life of 25,344 hours0.00398 days <br />0.0956 hours <br />5.687831e-4 weeks <br />1.30892e-4 months <br /> (see Table 4-2).

Input to the analysis is shown in Table 4-4. This analysis used the assump-tions on densification described in reference 2.

4.2.2. Cladding Stress The ANO Unit 1 fuel densification report 3 analysis indicated that the batch 1 fuel was the most limiting from a stress point of view. Batch 4 fuel has a higher nominal theoretical density and a greater pin prepressurization than the batch 1 fuel. Therefore, the analysis performed in the fuel densification 3

report conservatively envelopes the worst-case conditions for Cycle 2 fuel.

4.2.3. Fuel Pellet Irradiation Swelling The fuel design criteria sepcify a limit of 1.0% on cladding circumferential i

plastic strain.

The pellet design is such that the plastic cladding strain is less than 1% at 55,000 mwd /mtU. The conservatisms in this analysis are listed below.  !

1.

The maximum specification value for the fuel pellet diameter was used.

2.

The maximum specification value for the fuel pellet density was used.

3. )

The cladding ID used was the lowest permitted specification tolerance.

4.

The maximum expected three-cycle local pellet burnup is less than 55,000 mwd /mtU.

4.3. Thermal Design The core loading for Cycle 2 operation is shown in Figure 3-1. There are 56 fresh (batch 4) FAs and 121 once-burned (batches 2 and 3) FAs, which are l thermally and geometrically similar. However, batch 4 fuel has a lower nominal initial density, slightly different dimensions,.and a higher linear heat rate capability (Table 4-5). The heat rate capability of batches 2 and 3' varied '

l l 4-2 Babcock & Wilcox

0 3 on cn ecceably basis b:ccurs th2 fuel pallets vara resintered to incrocca tha initial nominal density to 96.0% TD." This resintering resulted in pellet

(~ diameter and density variations. As described in reference 4, some of the U} variations were large enough to effectively reduce the allowable heat rate to which the resintered fuel may be exposed while maintaining the desired safety margins. To minimize selective fuel loading, a design linear heat rate limit of 19.4 kW/ft was selected for the batch 2 and 3 fuel in the Cycle 2 core. The

! linear heat rate capability for batch 4 fuel is 20.15 kW/ft. Linear heat rate capabilities are based on centerline fuel melt and were established utilizing the TAFY-35 code with full fuel densification penalties.

4.3.1. Power Spike Model The power spike model used in this analysis is the same as that presented in BAW-100556 with modificatio'ns applied to gF and Fk as described in reference

7. These probabilities have been changed to reflect additional data from operating reactors that support a somewhat different approach and yield less seu re penalties due to power spikes. Fg was changed from 1.0 to 0.5. Fk was l changed from a Gaussian to a linear distribution reflecting a decreasing fre- ,

quency with increasing gap size. The maximum gap size versus axial position is shown in Figure 4-1, while the power spike factor versus axial position is shown in Figure 4-2. These calculated values are based on an inicial fuel density of 93.5% TD and an enrichment of 3.0 wt % uranium-235. The gap size and power factor for the fuel in the Cycle 2 core would be smaller because of the lower enrichment of the batch 4 fuel and the higher density of batch 2 and 3 fuel. Although the initial enrichment of batch 3 fuel in Cycle 1 was slight-ly greater (3.05%), applying these factors to the Cycle 2 thermal-hydraulic de-sign (section 6) yields conservative results.

4.3.2. Fuel Temperature Analysis Thermal analysis of the fuel rods assumed in-reactor fuel densification to 96.5% TD. The basis for the analysis is given in references 5 and 6 with the following modifications:

1. The code option for no restructuring o'f fuel has been used in this analysis in accordance with the NRC's interim evaluation of TAFY.

2. The calculated gap conductance was reduced by 25% in accordance with the NRC's interim evaluation of TAFY.

t G

4-3 Babcock s.Wilcox

, - - , , ~ - - - - , , --w-

p q During Cycla 2 opsrctica tha high:st roletiva acscably powsr 1svale cccurred in batch 3 fuel (Figures 3-1 and 5-1). The fuel temperature analysis docu-s mented in the ANO-1 fuel densification report 3 is based on batch 1 fuel, (the

,) limiting fuel for Cycle 1) and is conservative with respect to the fuel in the Cycle 2 core. The results of this analysis predict an average fuel temperature of 1330F at 5.79 kW/ft, which is shown in Table 4-5 as the average fuel temper-ature at nominal heat rate for batch 2 and 3 fuel. A re-analysis, based on betch 4 fuel specification parameters, results in a slightly lower predicted average fuel temperature, 1315F at 5.8 kW/ft, shown in Table 4-5 for batch 4 fuel. 30th analyses are based on BOC (beginning-of-cycle, zero burnup) con-ditions.

4.4. Material Design The chemicc* compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 4 fuel assemblies is identical to that of the pres-ent fuel.

4.5. Operating Experience The Mark B-4 FA does not constitute a departure from past design philosophy.

Its adequacy has been verified by the operating experience of the six B&W O. 177-FA plants. As of September 1, 1976, the operating experience shown in Table 4-1 has been amassed for the six 177-FA plants using the Mark B' fuel assembly.

Table 4-1. Operating Experience Cumulative net Current Maximum assembly electrical output, Reactor cycle burnup, mwd /mtU MWh Oconee 1 3 21,304 13,826,959 Oconee 2 2 19,020 8,687,947 Oconee 3 1 19,163 8,745,066 .

Three Mile Is. 1 2 20,777 10,105,588 Arkansas one 1 15,450 6,922,082 Rancho Seco 1 7,553 2,042,758

• i J

4-4 Babcock & Wilcox

0 O '

Tr:ble 4-2. Fut1 Design Para.oetera, Residual fuel New fuel assembly assembly Batch 2 Batch 3 Batch 4 Fuel assembly type Mark B-3 Mark B-3 Mark B-4 Number of assemblies 61 60 56 Initial fuel enrichment 2.72 3.05 2.64 Initial fuel density, % TD >93.11I ") >93.11(*) 93.5 Batch burnup, BOC, mwd /mtU 17,805 11,704 0 Design life, EFPH 18,288 25,344 20,640 Cladding collapse time, >30,000 >30,000 s30,000 (b)

EFPH

(* Variable due to additional sintering of batch 2 and 3 pellets.

( )A detailed three-cycle collapse analysis for batch 4 will be performed for the Cycle 3 reload report. A cladding collapse time of %30,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is a preliminary estimate based on a comparison of batch 4 design parameters with those of batches 2 and 3.

O G

l 4_3 Babcock & Wilcox

. _ . , , -_ . _ . . _ , . ~

(^ D I

Table 4-3. Fuel Rod Nominal Dimensions t

l l

t Residual New fuel s fuel assembly, assembly, i ) Component batches 2 and 3 batch 4 Fuel rods OD, in. 0.430 0.430 ID, in. 0.377 0.377 Fuel Pellet Diameter, in. 0.367 0.370 Density, % TD 96.0 93.5 Undensified Active Fuel Length, in. 141.0 142.6 Flexible spacers, type corrugated / spring spring Solid spacers, material Zr02 none Tubular spacers, material Zr02 Zr-4 I\ Table 4-4.

Input Summary for Cladding Creep

Collapse Calculations Residual fuel assembly, batches 2 and 3

1. Pellet diameter, in. 0.366

2. Pellet density, % TD 93.ll(a)

3. Densified pellet diameter, in. 0.364

4. Cladding ID, in. 0.377

5. Reactor system pressure, psia 2200

6. Stack height (undensified), in. 144.0 l

(a) Represents a conservative value for collapse analysis.  !

l i

-C\

\J l 4-6 Babcock & \Milcox .

l

p r.

T-blo 4-5. Full Tharmal Analysis Parameters Batches 2 and 3 Batch 4 Fuel pellet, nominal Initial density, % TD 96.0(*) 93.5 Initial stack length, in. 141.00(a) 142.6 Initial diameter, in. O.367(a) 0,370 Densified density, % TD 96.5 96.5 Densified stack length, in. 140.65 140.49 Densified diameter, in.

0.3646 0.3645 Hot channel factor on linear heat rate 1.014 1.014 Nominal linear heat rate, kW/fe 5.79(b) 5.80 (b)

Averagm fuel temperature at nominal . 1330(c) 1315 linear heat rate, F Linear heat rate to centerline fuel 19.40(d) 20.15(*)

me.lt, kW/ft I*} Nominal values after resintering. *

(b) Based on densified length.

O, (*) Based on batch 1 fuel, which is more limiting than either batch 2 or 3.

(d) Design limit. The linear heat rate capability of each fuel assembly in batches,2 and 3 has been determined from resinter data."

I*) Minimum capability based on batch 4 fuel specification.

i lO 4-7 Babcock & Wilcox

O O O Figure 4-1. Maximum Gap Size Vs Axial Position - ANO Unit 1 Cycle 2 3

Based on densification from 93.5 to 96.5% TD g2 -

i 5

o 0

31 -

E t

o>

0 1 I I I l I O 20 40 60 80 100 120 140 Axial Location, in.

J i

i e

O 8

x P

9. -

w